The challenge of producing fusion power is hugely complex. Fusion neutrons are produced when a deuterium-tritium (D-T) or deuterium-deuterium (D-D) plasma becomes very hot so that the nuclei fuse together, releasing energetic neutrons. To date, the most promising way of achieving this is to use a tokamak; in the conventional tokamak approach to fusion (as embodied by ITER), the plasma needs to have high confinement time, high temperature, and high density to optimise this process.
A tokamak features a combination of strong toroidal magnetic field BT, high plasma current IP and usually a large plasma volume and significant auxiliary heating, to provide a hot stable plasma so that fusion can occur. The auxiliary heating (for example via tens of megawatts of neutral beam injection of high energy H, D or T) is necessary to increase the temperature to the sufficiently high values required for nuclear fusion to occur, and/or to maintain the plasma current.
The problem is that, because of the large size, large magnetic fields, and high plasma currents generally required, build costs and running costs are high and the engineering has to be robust to cope with the large stored energies present, both in the magnet systems and in the plasma, which has a habit of ‘disrupting’—mega-ampere currents reducing to zero in a few thousandths of a second in a violent instability.
The situation can be improved by contracting the donut-shaped torus of a conventional tokamak to its limit, having the appearance of a cored apple—the ‘spherical’ tokamak (ST). The first realisation of this concept in the START tokamak at Culham demonstrated a huge increase in efficiency—the magnetic field required to contain a hot plasma can be reduced by a factor of 10. In addition, plasma stability is improved, and build costs reduced.
WO 2013/030554 describes a compact spherical tokamak for use as a neutron source or energy source. An important consideration in the design of spherical tokamaks is the protection of reactor components from the high neutron flux generated by the fusion reaction. This is of particular importance on small tokamaks as the neutron flux (i.e. neutron flow per unit area) will in general be higher due to the smaller surface area-to-volume ratio of the plasma vessel.
The present application is based on a very compact form of the tokamak, and employs a range of innovative features, including use of High Temperature Superconducting magnets. The ‘Efficient Compact Fusion Reactor’ (ECFR) is intended to provide a compact fusion power plant. FIG. 1 is a schematic diagram of such a reactor. The plasma (11) is contained within a vacuum vessel (12) by the magnetic fields generated by a toroidal field coil (13) and a poloidal field coil (not shown). The toroidal field coil runs down a central column (14) in the centre of the plasma chamber.
A drawback of the ST is that the limited space in the central column prohibits installation of the substantial shielding necessary to protect the central windings in a neutron environment—so conventional toroidal field windings, and conventional central solenoids (used to induce and maintain the plasma currents) are not practical. Although power plants based on the ST have been designed (using solid copper centre posts with limited shielding, the post being changed every year or so when damaged by neutrons), these have high energy dissipation in the centre column due to the relatively high resistivity of warm copper, requiring a large device for electricity production to become economical.
Superconducting materials may be used for the central core, but such materials are vulnerable to damage from neutrons, and may fail catastrophically if enough damage accumulates that the material no longer superconducts. There is therefore a trade-off between the overall size of the central core, the cross sectional area of the superconducting material (which is related to the maximum current that the superconductor can carry), and the thickness of the shielding.
In order to ensure that the reactor is as compact as possible (which allows greater efficiency), the thickness of shielding should be reduced as much as possible, while still maintaining adequate protection for the other components. Minimising the distance between the plasma and the field coils allows a higher magnetic field in the plasma with a lower current in the coils.
FIG. 2 shows a section of the central column, and illustrates the problems which the shielding material must overcome. The central column (13) comprises a central core of HTS coils (21) and an outer layer of shielding (22). Depending on the material used for the shielding, there may be a layer of oxidised shielding material (23) on the outer surface. There are three major causes of damage which originate from the plasma. Firstly, the high energy neutrons generated by the fusion reaction can essentially knock atoms out of the structure of the shielding, creating damage cascades which propagate through the material. Secondly, the heat flux from the fusion reaction is significant, and can damage the shielding due to thermal stresses induced by uneven heating and the HTS core, as higher temperatures reduces the current that can be carried while maintaining superconductivity, and can cause the coil to suddenly gain resistance, causing the magnet to quench. Lastly, the energetic particles of the plasma will ablate the outer surface of the shielding. This not only causes damage to the shielding itself, but can also contaminate the plasma. It is desirable to have a shielding material which can resist these effects, as well as prevent neutrons from reaching the superconducting coils.